JP2024012771A - Nuclear power plant reliability improvement method - Google Patents

Abstract

To provide a nuclear power plant reliability improvement method for improving the reliability of a nuclear power plant by performing efficient preventive maintenance for flow accelerated corrosion through application of maintenance with importance on reliability taking a corrosion potential as a main monitoring index.SOLUTION: A nuclear power plant reliability improvement method has: a selection step of selecting a piping part with a high risk of flow accelerated corrosion; a monitoring step of monitoring the corrosion rate of the flow accelerated corrosion at the selected piping part; a preventive maintenance execution step of reducing the corrosion rate at the piping part; a correction step of controlling a water quality parameter for cooling water flowing in the piping part; an improvement step of replacing the material of the selected piping part when the corrosion rate does not fall within a target range even if the correction step is performed; and a lifecycle management step of managing the corrosion rate of piping of the nuclear power plant through the selection step, monitoring step, preventive maintenance step, correction step, and improvement step.SELECTED DRAWING: Figure 1

Description

The present invention relates to a method for improving the reliability of a nuclear power plant.

Carbon steel is often used as a material for piping in thermal power plants and nuclear power plants. Carbon steel piping is known to suffer from flow-accelerated corrosion (FAC) under certain conditions where fluids flow at high flow rates. FAC occurs primarily due to flow rate dependent electrochemical corrosion. FAC precautions are applied because FAC results in a reduction in pipe wall thickness.

In the case of nuclear power plants, carbon steel is often used for ex-core piping installed outside the pressure vessel. At the plant design stage, corrosion allowances are set in advance for carbon steel piping through which high-velocity cooling water flows. The corrosion allowance is set in anticipation of the amount of wall thinning due to FAC during the plant's service life. During the service period, the amount of wall thinning due to corrosion is periodically inspected.

In recent years, for carbon steel piping, it is desired not only to provide a corrosion allowance in advance, but also to optimize preventive maintenance throughout the service life. Efficient countermeasures against FAC are required from the viewpoints of improving plant safety, improving economic efficiency such as capacity utilization, and coping with aging. Regarding carbon steel piping that is prone to FAC, it is desired to optimize the timing of inspection and replacement, and to ensure equipment reliability through low-cost and rational measures.

In the case of nuclear power plants, measures to suppress FAC include changing materials from carbon steel to corrosion-resistant materials such as low alloy steel and stainless steel. Additionally, in terms of water quality control, oxygen is injected into the cooling water. FAC occurs when dissolved oxygen concentration drops below 15-20 ppb. When the dissolved oxygen concentration decreases, it becomes difficult to form an oxide film on the inner surface of the pipe, and the protection of the base material is lost, making FAC more likely to occur.

In boiling water reactors (BWRs), oxygen injection into the water supply and condensate systems is widely practiced both domestically and internationally. Steam generated in nuclear reactors contains oxygen produced by radiolysis of water. However, when steam is condensed in a condenser, most of the oxygen remains in the gas phase. Therefore, the dissolved oxygen concentration of condensate often decreases to 10 ppb or less. Since an oxide film is less likely to be formed in the piping through which condensate flows, FAC is more likely to occur when high-temperature, high-flow rate cooling water flows. The FAC of such piping is countered by oxygen injection.

In a pressurized water reactor (PWR), oxygen is injected into the secondary system and pH is adjusted. The secondary cooling water is adjusted to be alkaline by adding ammonia or the like. It is known that on the alkaline side, the protection of the base material by the oxide film is less likely to be lost, and FAC is less likely to proceed even if the dissolved oxygen concentration is reduced to about 10 ppb by addition of hydrazine or deaeration.

Dissolved oxygen concentration, pH, and iron concentration of cooling water are mainly used as indicators for water quality management regarding FAC. Water quality management targets include water supply, condensate, reactor water, steam condensed water, and suppression pool water. Regarding the piping system through which these cooling waters flow, the amount of wall thinning due to FAC is calculated based on consideration of the flow rate of the cooling water, the temperature of the cooling water, the pipe diameter, and the geometric shape of the pipes such as elbows and vents. is managed.

Conventionally, in some boiling water reactors (BWRs), hydrogen injection or noble metal injection has been carried out as a measure against stress corrosion cracking (SCC) of materials that come into contact with cooling water. As an index for water quality management regarding SCC, electrochemical corrosion potential (ECP) is used as a higher index than the dissolved oxygen concentration of cooling water. It is known that stainless steel becomes less susceptible to SCC when the ECP is lower than about -300 to -200 mV (vs. SHE).

Conventionally, in nuclear power plants, various countermeasures have been taken against various corrosion phenomena such as SCC and erosion/corrosion based on evaluation of corrosion rate.

Patent Document 1 describes a nuclear power plant management method for managing the occurrence of stress corrosion cracking in structural members that come into contact with cooling water of the nuclear power plant. This method measures the ECP of structural members that come into contact with cooling water and the concentration of impurity ions contained in the cooling water. Based on this information, the time required for stress corrosion cracking to occur is determined. When the generation time is shorter than the set time, at least one of the ECP and impurity ion concentrations is reduced.

Patent Document 2 describes a thinning evaluation system for piping systems in power generation plants and the like. This system constructs a mathematical formula that expresses the rate of wall thinning using the measurement results of pipe temperature, pipe internal fluid temperature, pipe fluid wetness, and pipe fluid velocity as parameters. are doing. Based on the maximum wall thinning rate in the piping system, the inspection interval, remaining life of the piping, and annual plan schedule for wall thickness measurement are determined.

Patent Document 3 describes a method for calculating and evaluating wall thinning due to erosion and corrosion of equipment and piping devices. In this method, an online thinning monitoring system is configured for each system of the actual plant. Based on the thinning data from actual plants, we added thinning measurement data from actual plants, corrected functional formulas, added and modified the database of thinning calculation formulas, and ranked and weighted erosion and corrosion factors.

Patent No. 5483385 Japanese Patent Application Publication No. 2006-138480 Japanese Patent Application Publication No. 8-178172

In conventional nuclear power plants, measures to prevent pipe corrosion have generally been taken individually. For example, in countermeasures against stress corrosion cracking (SCC), suppression of the progression of SCC, replacement of piping where SCC was discovered, and confirmation of the soundness of replaced piping were carried out individually for each piping section. . When operating a plant while allowing SCC, the amount of progress has been evaluated and SCC that has already occurred has to be continuously inspected.

The same applies to flow-accelerated corrosion (FAC), and countermeasures have been taken for each individual piping section. At the plant design stage, corrosion allowances were set for specific piping to account for the amount of wall thinning caused by FAC during the service period. The amount of thinning due to corrosion was individually and periodically inspected. The standard FAC management method was to predict and manage future wall thinning based on countermeasures and inspections for each piping section.

However, the conventional FAC management method has a problem in that the number of piping parts to be managed becomes enormous. From the viewpoint of preventive maintenance of FAC in the entire plant, it is necessary to periodically inspect a certain percentage of piping parts where FAC may occur. Periodic inspections must be conducted to ensure that the amount of wall thinning expected to occur within a specified number of years is within the design range. If the number and range of objects to be managed are enormous, problems such as inspection costs, plant operating rates, and human errors will arise.

Patent Document 1 describes a technique that uses the measurement results of the corrosion potential (ECP) of a material to estimate the time at which stress corrosion cracking (SCC) occurs. However, the ECP measurement results are used for inputting and correcting model equations. The accuracy of the model formula needs to be confirmed by comparing it with actual measurement test results. Conventional techniques still have the problem that the number of objects to be managed is enormous.

If the corrosion rate is estimated based on the model equation, it becomes possible to formulate future inspection plans. However, when the number of objects to be managed is enormous, the time interval between periodic inspections becomes longer for a certain piping section. There is no guarantee that corrosion will not occur beyond the design range by the next inspection. Estimation based on the model formula alone does not provide sufficient validity for extending the time interval between inspections for rationalization.

In recent years, the introduction of Reliability Centered Maintenance (RCM) has been promoted worldwide in nuclear power plants. A typical document regarding RCM is AP913 of the Institute of Nuclear Power Operations (INPO). RCM selects important structures, systems and components (SSC), optimal maintenance methods, and implementation timing for maintenance measures in order to ensure equipment reliability (ER). , develop an optimal maintenance program.

In the future, the introduction of RCM will become important in nuclear power plants from the viewpoints of reducing inspection costs, improving plant availability, reducing human errors, etc. The introduction of RCM requires SSC ER throughout the plant. However, at present, there is no routine ER process based on RCM regarding FAC, which affects the reliability of nuclear power plants. The mainstream of conventional measures to prevent FAC is Condition Base Maintenance (CBM), which is only applied individually depending on the content of the measure or the object of management.

RCM requires constant monitoring of certain performance indicators to indicate the status of the ER. However, with regard to FAC, there is no standard ER process based on RCM, so there is no established performance index to constantly monitor. If there is an index with high validity as a constantly monitored performance index, it will be possible to ensure the validity of the estimation result when the corrosion rate is estimated by FAC. Even if the periodic inspection time interval is extended, SSC ER becomes possible, so it is expected that preventive maintenance for FAC with streamlined inspections will be realized.

Therefore, the present invention aims to improve the reliability of nuclear power plants by performing efficient preventive maintenance against flow accelerated corrosion (FAC) by applying reliability-oriented maintenance (RCM) with corrosion potential (ECP) as the main monitoring index. The purpose of this study is to provide a method for improving the reliability of nuclear power plants.

In order to solve the above problems, the method for improving the reliability of a nuclear power plant according to the present invention includes a selection step of selecting piping parts with a high risk of flow-accelerated corrosion, and a selection step of selecting piping parts with a high risk of flow-accelerated corrosion of the selected piping parts. a monitoring step of monitoring a corrosion rate; a step of performing preventive maintenance to reduce the corrosion rate of the piping section; a corrective step of controlling water quality parameters of cooling water flowing through the piping section; When the corrosion rate is not within the target range, the improvement step of replacing the material of the selected pipe section, the selection step, the monitoring step, the preventive maintenance execution step, the correction step, and the improvement step, a life cycle management step of managing the corrosion rate of plant piping.

According to the present invention, by applying reliability-oriented maintenance (RCM) with corrosion potential (ECP) as the main monitoring index, the reliability of nuclear power plants is improved by performing efficient preventive maintenance against flow-accelerated corrosion (FAC). A method for improving the reliability of a nuclear power plant can be provided.

FIG. 1 is a block diagram showing a method for improving reliability of a nuclear power plant according to an embodiment of the present invention. 1 is a diagram showing an example of a nuclear power plant to which a method for improving reliability of a nuclear power plant according to an embodiment of the present invention is applied. 1 is a diagram showing an example of a reactor coolant purification system of a nuclear power plant. FIG. 2 is a diagram showing a specific example of a flow of a method for improving reliability of a nuclear power plant according to an embodiment of the present invention. FIG. 2 is a diagram illustrating an example of application of a flow of a method for improving reliability of a nuclear power plant according to an embodiment of the present invention to a BWR. 1 is a diagram showing an example of a nuclear power plant to which a method for improving reliability of a nuclear power plant according to an embodiment of the present invention is applied.

Hereinafter, a method for improving the reliability of a nuclear power plant according to an embodiment of the present invention will be described. Note that common components in the following figures are denoted by the same reference numerals, and redundant explanation will be omitted.

≪Methods for improving reliability of nuclear power plants≫
The method for improving the reliability of a nuclear power plant according to the present embodiment relates to a method for ensuring the reliability of structures, systems, and equipment (SSC) against flow accelerated corrosion (FAC) by applying reliability-oriented maintenance (RCM). . The monitoring targets that are constantly monitored in RCM are piping parts that are predicted to have a high risk of FAC. The constantly monitored performance index may be the corrosion potential (ECP) to be monitored. Although the wall thickness of the piping portion can be used as a performance index, it is preferable to use ECP.

In this specification, constant monitoring means to continuously collect information through measurement to the extent that trends in the measurement target can be grasped over most of the time range of the measurement cycle. Continuous monitoring may include, as part of a measurement cycle performed at arbitrary time intervals, a short measurement stop period for replacement of measurement equipment or the like.

The method for improving the reliability of a nuclear power plant according to this embodiment includes measuring performance indicators such as ECP to be monitored, estimating the corrosion rate of FAC (FAC rate) based on the measurement results of performance indicators such as ECP, and Actions for the FAC based on the determined FAC speed are implemented throughout the life cycle of the plant to realize the ER process for the FAC. By applying RCM, it is possible to optimize measures for FAC, to optimize targets for periodic inspections involving actual measurements, and to optimize time intervals for periodic inspections. These optimizations provide efficient preventive maintenance for FACs with streamlined inspections and improve nuclear plant reliability.

As the piping part to be monitored, an appropriate section of the piping of a nuclear power plant can be adopted. The piping portion to be monitored may be any of a section made up of a single pipe, a section made up of a part of a single pipe, and a section made up of a plurality of pipes connected to each other. Measures for FAC can be targeted at the section including the piping part to be monitored.

FIG. 1 is a block diagram showing a method for improving reliability of a nuclear power plant according to an embodiment of the present invention.
As shown in FIG. 1, the method for improving the reliability of a nuclear power plant according to the present embodiment includes a selection step S1, a monitoring step S2, a preventive maintenance execution step S3, a correction step S4, an improvement step S5, and a life cycle. management step S6. FIG. 1 shows newly devised monitoring indicators and each step for FAC based on the diagram of INPO's AP913.

The selection step S1 is a step of classifying devices or parts and selecting important elements as monitoring targets. In the selection step S1, piping parts with a high risk of flow accelerated corrosion (FAC) in a nuclear power plant are selected.

The monitoring step S2 is a step of monitoring the performance of the system and elements. In the monitoring step S2, performance indicators such as ECP of the piping part selected as the monitoring target are monitored, and the corrosion rate (FAC rate) of flow accelerated corrosion (FAC) of the piping part is estimated based on the performance index. Monitor.

The preventive maintenance execution step S3 is a step for executing preventive maintenance measures regarding equipment conditions and the like. In the preventive maintenance execution step S3, preventive maintenance measures are executed to reduce the flow accelerated corrosion (FAC) corrosion rate (FAC rate) of the piping portion selected as the monitoring target.

The corrective step S4 is a step for implementing corrective measures such as improvement and maintenance. In the corrective step S4, corrective measures are executed to control the water quality parameters of the cooling water flowing through the piping portion selected as the monitoring target.

The improvement step S5 is a step of implementing improvement measures to continuously improve device reliability. In the improvement step S5, improvement measures are performed to replace the material of the piping portion selected as the monitoring target.

The life cycle management step S6 is a step for managing life extension and the like through monitoring of systems and elements. In the life cycle management step S6, by repeating the selection step S1, the monitoring step S2, the preventive maintenance execution step S3, the correction step S4, and the improvement step S5, the corrosion rate (FAC rate) of flow accelerated corrosion (FAC) of the piping of the nuclear power plant is determined. ).

In steps S3 to S5, various FAC measures are taken for the piping portion selected as a monitoring target in step S1 so that the FAC speed falls within the target range. By executing steps S3 to S5, the risk of FAC of the piping site selected in step S1 is reduced. As a result, the ranking of FAC risks changes among the piping parts within the plant. In step S6, in accordance with such changes in the ranking, the piping parts to be monitored are reselected, and the performance indicators of the monitoring target are measured, the FAC speed is estimated based on the measurement results of the performance indicators, and the measures for the FAC are taken. repeat.

Note that the monitoring step S2 can be executed at an appropriate stage. For example, the monitoring step S2 can be arbitrarily executed after the selection step S1, the preventive maintenance execution step S3, the correction step S4, the improvement step S5, etc. Depending on the result of the monitoring step S2 after the execution of each step, it is possible to select the next step and determine whether or not the step should be executed.

According to the method for improving the reliability of a nuclear power plant according to the present embodiment, performance indicators such as ECP are constantly monitored with high-risk piping parts of FAC as monitoring targets, so monitoring is estimated based on performance indicators such as ECP. A subject's FAC speed can be managed to fall within a target range. Therefore, it is possible to reduce the FAC risk of the entire structure, system, and equipment (SSC) of a nuclear power plant to a level that allows preventive maintenance to be reasonably practicable (As Low As Reasonably Practicable: ALARP). In configuration management, it is shown that each SSC is in the desired state as designed and is exhibiting the desired performance, so equipment reliability with respect to the plant's FAC is good. can be clearly indicated.

Therefore, according to the method for improving the reliability of a nuclear power plant according to the present embodiment, by applying RCM, it is possible to perform efficient preventive maintenance on the FAC and improve the reliability of the nuclear power plant. Individual measures to prevent FAC can be coordinated with overall management procedures through the introduction of RCM. Such collaboration makes it possible to create a process that continuously improves the reliability of the entire nuclear plant over the long term and throughout its lifecycle. Conventionally, it was necessary to confirm the amount of pipe wall thinning due to FAC through periodic inspections involving actual measurements of a huge number of pipe parts. On the other hand, according to the application of RCM, from a long-term perspective, part of the periodic inspection of the amount of thinning by FAC can be replaced by constant monitoring of monitoring indicators such as ECP. Since it is possible to extend the time interval between inspections, reduce the frequency of inspections, and reduce the number of parts to be inspected, it is possible to streamline inspections and reduce inspection costs while ensuring equipment reliability.

The method for improving the reliability of a nuclear power plant according to this embodiment can be applied to any appropriate type of nuclear power plant. Types of nuclear power plants include boiling water reactors (BWR), advanced boiling water reactors (ABWR), pressurized water reactors (PWR), natural uranium graphite moderated carbon dioxide cooled reactors (GCR), and improved Examples include a type gas cooled reactor (AGR), a high temperature gas reactor (HTR), a heavy water reactor (HWR), a molten salt reactor (MSR), and a fast breeder reactor (FBR).

≪Configuration example of nuclear power plant≫
FIG. 2 is a diagram showing an example of a nuclear power plant to which a method for improving reliability of a nuclear power plant according to an embodiment of the present invention is applied. FIG. 2 illustrates a boiling water reactor (BWR).
As shown in FIG. 2, the nuclear power plant P100 includes a reactor P1, a turbine P3, a condenser P4, a reactor coolant purification system, a water supply system, and the like. The nuclear power plant P100 is configured to be able to perform a combination of hydrogen injection and noble metal injection during operation.

The nuclear reactor P1 is stored inside a containment vessel P11. The nuclear reactor P1 includes a pressure vessel P12. A reactor core P13 is contained inside the pressure vessel P12, and a cylindrical shroud P15 is installed so as to surround the reactor core P13. The shroud P15 is supported inside the pressure vessel P12 by a shroud support P41 attached to the inside of the pressure vessel P12.

A plurality of fuel assemblies are loaded into the core P13. A fuel assembly is formed by accommodating a plurality of fuel rods in a lattice shape inside a channel box. A fuel rod contains a plurality of fuel pellets made of nuclear fuel material inside a cladding tube. A neutron instrumentation tube P38 for measuring neutron flux is inserted into the reactor core P13.

An annular downcomer P17 is formed between the inner surface of the pressure vessel P12 and the outer surface of the shroud P15. A plurality of jet pumps P21 are installed at the bottom of the pressure vessel P12. The outlet of the jet pump P21 is located at the bottom of the downcomer P17. The jet pump P21 discharges the cooling water present in the downcomer P17 of the pressure vessel P12 to a lower plenum, which is a space below the pressure vessel P12.

A water supply system is connected to the pressure vessel P12. The water supply system includes a water supply pipe P10, a water supply pump P5, a condensate purification device P6, a water supply pump P7, a low pressure water heater P8, a high pressure water heater P9, and the like. These devices are installed in this order on the water supply pipe P10. The water supply pipe P10 connects the condenser P4 to the pressure vessel P12 through a pipe line.

A hydrogen injection device P16 is connected to the water supply system via a hydrogen injection pipe P18. The hydrogen injection pipe P18 is connected to a section of the water supply pipe P10 between the condensate purification device P6 and the water supply pump P7. The hydrogen injection pipe P18 is provided with an on-off valve P19 that can be opened and closed. The hydrogen injection device P16 is a device for injecting hydrogen, and injects hydrogen gas into the cooling water supplied to the nuclear reactor P1.

A noble metal injection device P31 is connected to the water supply system via a noble metal injection pipe P32. The noble metal injection pipe P32 is connected to a section of the water supply pipe P10 between the high pressure water heater P9 and the pressure vessel P12. The noble metal injection pipe P32 is provided with an on-off valve P33 that can be opened and closed. The noble metal injection device P31 is a device for performing noble metal injection, and injects a solution of a noble metal compound into the cooling water supplied to the nuclear reactor P1.

A reactor coolant purification system is connected to the pressure vessel P12. The purification system is composed of a bottom drain pipe P34, a purification system pipe P20, a purification system isolation valve P23, a regenerative heat exchanger P25, a non-regenerative heat exchanger P26, a purification system pump P24, a reactor water purification device P27, etc. . These devices are installed in this order on the purification system piping P20. The bottom drain pipe P34 is connected to the bottom of the pressure vessel P12. The purification system piping P20 connects from one end of the bottom drain piping P34 to the section of the water supply piping P10 between the high pressure water heater P9 and the pressure vessel P12.

An oxygen injection device P43 is connected to the purification system via an oxygen injection pipe P44. The oxygen injection pipe P44 is connected to a section of the purification system pipe P20 between the pressure vessel P12 and the regenerative heat exchanger P25. The oxygen injection pipe P44 is provided with an on-off valve P45 that can be opened and closed. The oxygen injection device P43 is a device for injecting oxygen or hydrogen peroxide, and injects oxygen gas or an aqueous solution of hydrogen peroxide into the cooling water extracted from the reactor P1.

A recirculation system is connected to the pressure vessel P12. The recirculation system includes a recirculation system piping P30, a recirculation system pump P37, and the like. The recirculation system piping P30 connects the lower part of the downcomer P17 of the pressure vessel P12 to the upper part of the downcomer P17 via the outside of the pressure vessel P12. The recirculation system piping P30 branches upstream of the recirculation system pump P37. The branched pipe joins the purification system pipe P20 via an on-off valve P23 that can be opened and closed.

A corrosion potential sensor P35a is installed in the purification system. The corrosion potential sensor P35a is installed in a branch pipe P34a branched from the bottom drain pipe P34. The corrosion potential sensor P35a is attached to a flange P36 connected to the branch pipe P34a. The branch pipe P34a is connected to a cooling water sampling line or the like.

A corrosion potential sensor P35b is installed in the recirculation system. The corrosion potential sensor P35b is installed in the recirculation system piping P30. The corrosion potential sensor P35b is attached to a flange P36 connected to the recirculation system piping P30.

The corrosion potential sensors P35a and P35b measure the corrosion potential (ECP) of the material that comes into contact with the cooling water under the quality of the cooling water extracted from the pressure vessel P12. The corrosion potential sensors P35a and P35b include a working electrode that measures the potential of the material and a reference electrode that generates a reference potential in cooling water. Corrosion potential sensors P35a and P35b are provided to confirm the effect of suppressing stress corrosion cracking (SCC) by hydrogen injection or noble metal injection.

Note that a corrosion potential sensor for checking the effect of suppressing SCC can also be installed inside the pressure vessel P12. In FIG. 2, a corrosion potential sensor P35c is installed in the lower plenum. A corrosion potential sensor P35d is installed in the core 13. These corrosion potential sensors P35c and P35d measure the corrosion potential (ECP) of the material that comes into contact with the cooling water under the quality of the cooling water inside the pressure vessel P12.

In the nuclear power plant P100 shown in FIG. 2, cooling water is supplied into the pressure vessel P12 from the water supply pipe P10 of the water supply system. Cooling water is ejected from a water supply sparger above the downcomer P17, and is supplied to the core P13 by a jet pump P21. The cooling water is heated in the reactor core P13 by heat generated by fission of the nuclear fuel material. The heated gas-liquid two-phase cooling water is separated into steam and water in a steam-water separator.

The water separated by the steam separator descends down the downcomer P17 and returns to cooling water. On the other hand, the steam separated by the steam separator is sent to the turbine P3 through the main steam pipe P2 after moisture is removed by the steam dryer. Turbine P3 is connected to a generator. Electric power is generated by the rotation of the turbine P3 by the steam. The steam that has consumed energy in the turbine P3 is cooled in the condenser P4 and returns to cooling water.

Cooling water is supplied from the condenser P4 toward the pressure vessel 12 through a water supply pipe P10. After being discharged from the condenser P4, the cooling water is pressurized by the condensate pump P5 and introduced into the condensate purification device P6. The condensate purifier P6 removes impurities contained in the cooling water. The purified cooling water is pressurized by the feedwater pump P7, enters the low-pressure feedwater heater P8, and is then introduced into the high-pressure feedwater heater P9. In the low-pressure feedwater heater P8 and the high-pressure feedwater heater P9, the cooling water is heated in stages to a temperature suitable for the thermal efficiency of the plant.

The turbine P3 and the high-pressure feed water heater P9 are connected to each other by a bleed pipe P14. The extracted steam extracted from the steam supplied to the turbine P3 bypasses the condenser P4, is introduced into the high pressure feed water heater P9, and is then introduced into the low pressure feed water heater P8. Thereafter, it joins the water supply pipe P10 upstream of the condensate pump P5. In the low pressure feed water heater P8 and the high pressure feed water heater P9, the cooling water is heated in stages using the heat of the extracted steam.

The cooling water inside the pressure vessel P12 contains metal corrosion products mixed in during water supply, metal corrosion products generated by corrosion of structural materials of the pressure vessel P12, and the like. Therefore, the cooling water inside the pressure vessel P12 is extracted at a constant rate and purified by the reactor coolant purification system. The cooling water inside the pressure vessel P12 is sucked from the lower side of the pressure vessel P12 toward the purification system piping P20 by the purification system pump P24.

The cooling water extracted from the pressure vessel P12 is introduced into the regenerative heat exchanger P25 and then into the non-regenerative heat exchanger P26. In the regenerative heat exchanger P25 and the non-regenerative heat exchanger P26, the cooling water is cooled to about 50° C., which is suitable for purification. The cooled cooling water is introduced into the reactor water purification device P27. The reactor water purification device P27 removes impurities such as metal corrosion products contained in the cooling water. The purified cooling water is introduced into the regenerative heat exchanger P25. In the regenerative heat exchanger P25, the cooling water is heated to a temperature suitable for the thermal efficiency of the plant.

When the operation of the nuclear reactor P1 is stopped, all control rods are inserted into the reactor core P13. Insertion of all control rods stops the fission chain reaction of nuclear fuel material. Heat remaining in the equipment inside the reactor core P13 and the pressure vessel P12 is removed by evaporation of cooling water. However, when the temperature of the cooling water decreases to a certain degree, the heat removal efficiency due to evaporation decreases. Therefore, when the temperature of the cooling water reaches about 150°C, the residual heat removal system is operated. The residual heat removal system cools the core structure and equipment by exchanging heat with the pressure suppression pool water and spraying cooling water.

In the nuclear power plant P100, hydrogen injection or a combination of hydrogen injection and noble metal injection is performed as a measure to suppress stress corrosion cracking (SCC) of materials that come into contact with cooling water. In the hydrogen injection, hydrogen gas is injected into the cooling water supplied to the nuclear reactor P1 by the hydrogen injection device P16. In the noble metal injection, a noble metal compound solution is injected into the cooling water supplied to the nuclear reactor P1 by the noble metal injection device P31.

Factors involved in the occurrence of SCC include mechanical factors such as tensile stress applied to the material, environmental factors to which the material is exposed, and material factors such as chemical components of the material. In cooling water, oxygen and hydrogen peroxide may be generated due to water radiolysis. Oxygen and hydrogen peroxide are environmental factors for SCC, and the higher the concentration, the more likely the SCC will develop. Therefore, in the nuclear power plant P100, hydrogen injection and noble metal injection are performed during operation in order to suppress SCC of materials that come into contact with cooling water.

Hydrogen injection is a technology in which hydrogen gas is injected into cooling water, causing the oxygen or hydrogen peroxide in the cooling water to react with the hydrogen and returning it to water. The hydrogen injected into the water supply system is ejected from the water supply sparger together with the cooling water above the downcomer P17, and mixes with the reactor water inside the pressure vessel P12. Since the dose rate of gamma rays around the downcomer P17 is moderate, the recombination reaction between oxygen and hydrogen peroxide and hydrogen is promoted. The recombination reaction consumes oxygen and hydrogen peroxide, which alleviates the corrosive environment.

Noble metal injection is a technique in which a solution of a noble metal compound is injected into cooling water to cause the noble metal to adhere to the surface of a material that comes into contact with the cooling water. As the noble metal compound, a platinum group compound such as sodium hexahydroxoplatinate is injected. The solution of the platinum group compound becomes a colloidal solution of the platinum group oxide by irradiation with gamma rays, and the platinum group element is attached to the surface of the material that comes into contact with the cooling water. Since platinum group elements catalyze recombination reactions, the corrosive environment can be alleviated with a small amount of hydrogen injection.

When hydrogen or noble metal is injected, the corrosion potential (ECP) of the material that comes into contact with the cooling water decreases. In the case of noble metal implantation, if hydrogen is present at twice the stoichiometric ratio of water to oxygen (H:O=2:1), the recombination reaction will proceed at around -500 mV (vs. SHE). do. Due to the combination of recombination reaction and material corrosion, the potential decreases to around -500 mV (vs. SHE). Hydrogen injection or noble metal injection reduces the ECP and suppresses the occurrence and progression of SCC.

When hydrogen is not injected, the dissolved oxygen concentration in the cooling water is usually on the order of several hundred ppb. On the other hand, when hydrogen is injected, the dissolved oxygen concentration of the cooling water decreases to about several ppb. A reduction in dissolved oxygen concentration will reduce the SCC susceptibility of materials that come into contact with cooling water. However, when hydrogen or noble metal is implanted, the dissolved oxygen concentration decreases, making it easier for FAC to proceed. FAC significantly thins carbon steel piping. Carbon steel piping is mainly used in the section outside the pressure vessel P12.

FIG. 3 is a diagram showing an example of a reactor coolant purification system of a nuclear power plant. FIG. 3 shows an example of the configuration of the reactor coolant purification system of the BWR shown in FIG. 2.
As shown in FIG. 3, in the purification system of the nuclear power plant P100, the regenerative heat exchanger P25 and the non-regenerative heat exchanger P26 are each configured with multiple stages of heat exchangers.

The regenerative heat exchanger P25 includes three stages of heat exchangers P25a, P25b, and P25c. The non-regenerative heat exchanger P26 includes two stages of heat exchangers P26a and P26b. The heat exchangers are connected to each other via communication pipes P42a, P42b, P42c, P42d, and P42e. Generally, the regenerative heat exchanger P25 has three stages, and the non-regenerative heat exchanger P26 has two stages. However, the number of stages of the heat exchanger is not particularly limited.

Cooling water extracted from the reactor P1 is sent to the regenerative heat exchanger P25 through the purification system piping P20. The cooling water is cooled by heat exchange with the cooling water purified by the reactor water purification device P27 in the heat exchangers P25a, P25b, and P25c of each stage. Thereafter, it is introduced into the non-regenerative heat exchanger P26 through the communication pipe P42d. The cooling water is cooled by heat exchange with the auxiliary equipment cooling water in the heat exchangers P26a and P26b of each stage. Thereafter, the pressure is increased by the purification system pump P24 and introduced into the reactor water purification system P27.

After the cooling water is cooled by the regenerative heat exchanger P25 and the non-regenerative heat exchanger P26, impurities are removed by the reactor water purification device P27. The purified cooling water is sent to the regenerative heat exchanger P25. The cooling water is heated by heat exchange with the cooling water extracted from the reactor P1 in the heat exchangers P25a, P25b, and P25c of each stage of the regenerative heat exchanger P25. Thereafter, the water is sent to the water supply pipe P10 and is supplied to the reactor P1 together with the condensate.

In the regenerative heat exchanger P25 and the non-regenerative heat exchanger P26, the cooling water extracted from the reactor P1 is cooled to a temperature suitable for purification. Furthermore, in the regenerative heat exchanger P25, the cooling water purified by the reactor water purification device P27 is heated to a temperature suitable for the thermal efficiency of the plant. Auxiliary cooling water is supplied to the non-regenerative heat exchanger P26 from the auxiliary cooling system. The auxiliary cooling water circulates through the auxiliary cooling system passing through each stage of the non-regenerative heat exchanger P26, and is cooled by heat exchange with seawater or the like.

Flow accelerated corrosion (FAC) is an electrochemical corrosion involving mass transfer due to dissolution of an oxide film. In addition to the flow rate of the cooling water, the FAC depends on the chemical composition of the piping section, the quality of the cooling water, the temperature of the cooling water, the hydrodynamic properties of the cooling water at the piping section, and the like. The FAC speed increases as the flow rate of cooling water increases. Furthermore, the higher the temperature of the cooling water, the faster the cooling rate, and generally reaches a maximum value around 130 to 150°C. Further, the speed increases when the dissolved oxygen concentration of the cooling water becomes about 15 ppb or less.

During operation of the nuclear power plant P100, the temperature of the cooling water extracted from the pressure vessel P12 is as high as about 280°C. The temperature of the cooling water extracted to the purification system decreases each time it passes through each stage of the regenerative heat exchanger P25 and the non-regenerative heat exchanger P26. On the downstream side of the non-regenerative heat exchanger P26, the temperature of the cooling water is 80° C. or lower. However, the temperature is relatively high on the upstream and downstream sides of the purification system. Further, the flow velocity in the purification system is high, on the order of several m/s, because the pipe diameter is small.

Further, when hydrogen injection or noble metal injection is performed during operation of the nuclear power plant P100, the oxygen concentration and hydrogen peroxide concentration of the cooling water also decrease in the purification system. In particular, when noble metals are implanted, excessive hydrogen tends to be generated, so that the amount of hydrogen with respect to the amount of oxygen tends to exceed the stoichiometric ratio of water. Oxygen and hydrogen peroxide in the cooling water may be completely consumed by recombination reactions before reaching the downstream side of the purification system.

Furthermore, the sections of the communication pipes P42a, P42b, P42c, P42d, and P42e and the purification system piping P20 outside the pressure vessel P12 are generally made of carbon steel. These piping may have structures that affect flow rates. Structures that affect flow rate include elbows, vents, tees, orifices, valves, and the like.

Piping that constitutes the purification system, especially the heating side communication pipe P42a that connects the first stage heat exchanger P25a and the second stage heat exchanger P25b of the regenerative heat exchanger P25, and the second stage heat exchanger P42a. A connecting pipe P42b on the heating side connects the heat exchanger P25b and the third stage heat exchanger P25c, and connects the second stage heat exchanger P25b and the first stage heat exchanger P25a. The cooling-side communication pipe P42f is made of carbon steel and is exposed to high flow rates and high temperatures, so it becomes a piping part with a high risk of FAC when the oxygen concentration of the cooling water decreases.

In the nuclear power plant P100, such piping parts with a high risk of FAC can be selected as monitoring targets, and oxygen injection or hydrogen peroxide injection can be performed as a measure to suppress the monitored FAC. Oxygen injection and hydrogen peroxide injection, like hydrogen injection and noble metal injection, can be performed during plant operation.

Oxygen injection is a technique that increases the oxygen concentration of cooling water by injecting oxygen gas into cooling water. Hydrogen peroxide injection is a technique in which an aqueous solution of hydrogen peroxide is injected into cooling water to increase the hydrogen peroxide concentration in the cooling water. When the concentration of the oxidizing agent in the cooling water is increased, an oxide film is formed on the surface of the material that comes into contact with the cooling water.

For example, in the case of carbon steel, hematite (Fe ₂ O ₃ ) can be formed. Hematite has a lower solubility than magnetite (Fe ₃ O ₄ ) and forms a dense oxide film. Increasing the concentration of the oxidant in the cooling water allows the oxide film to remain stable even if parameters related to mass transfer such as flow rate increase, thereby suppressing the FAC rate of materials that come into contact with the cooling water. Can be done.

In FIG. 3, a corrosion potential sensor P35f is installed in a communication pipe P42a on the heat radiation side that connects the first stage heat exchanger P25a and the second stage heat exchanger P25b of the regenerative heat exchanger P25. There is. Further, a corrosion potential sensor P35g is installed in a communication pipe P42f on the heat receiving side that connects the second stage heat exchanger P25b and the first stage heat exchanger P25a of the regenerative heat exchanger P25.

The corrosion potential sensors P35f and P35g are sensors for measuring the corrosion potential (ECP) of materials that come into contact with the cooling water. The corrosion potential sensors P35f and P35g include a working electrode that measures the potential of the material and a reference electrode that generates a reference potential in cooling water. Corrosion potential sensors P35f and P35g can be provided to monitor the FAC speed of communication pipes P42a and P42f.

≪ER process for FAC≫
Next, an explanation will be given of an equipment reliability ensuring (ER) process for flow accelerated corrosion (FAC) carried out in the nuclear power plant reliability improvement method according to the present embodiment. Note that in the following description, the application of the ER process for FAC to BWR will be exemplified. However, similar application is possible to PWR and other types by appropriately selecting the monitoring target.

The ER process for FAC can be started from any step among steps S1 to S5 shown in FIG. Normally, in order to specify a monitoring target to be constantly monitored, the process starts from a selection step S1.

(Selection step S1)
In the selection step S1, piping parts with a high risk of flow accelerated corrosion (FAC) in a nuclear power plant are selected. In this step, a piping system with a high risk of FAC is selected as a monitoring target to be constantly monitored. Then, among the piping parts included in the piping system, piping parts with a high risk of FAC are selected.

Selection of piping systems and piping locations is performed based on parameters that affect FAC and importance classification for structures, systems, and equipment (SSC). Based on these, piping systems and piping parts are ranked according to FAC risk. Then, a piping system with a high risk of FAC and a piping part included in the piping system with the highest risk of FAC are selected as monitoring targets to be constantly monitored.

Parameters that affect FAC include the flow rate of cooling water, the chemical composition of the piping site, the quality of the cooling water, the temperature of the cooling water, and the hydrodynamic properties of the cooling water at the piping site. The quality of the cooling water includes oxygen concentration, hydrogen peroxide concentration, electrical conductivity, pH, dissolved hydrogen concentration, and impurity concentration such as iron ions.

The importance classification is defined by classifying the importance of SSCs having safety functions. In terms of importance classification, the SSC of a nuclear power plant is classified into an abnormality prevention system (PS) and an abnormality mitigation system (MS). PS and MS are each classified into multiple classes 1 to 3 depending on the importance of safety functions. Based on the importance classification, it is possible to prioritize monitoring targets from the standpoint of nuclear plant safety.

Selection of piping systems and piping locations can be performed using risk analysis tools, human-induced risk analysis, or both. Examples of risk analysis tools include tools that quantify risks and automatically perform quantitative analysis. Human-induced risk analysis includes analysis based on meetings of experts, practitioners, and groups thereof, empirical analysis, deductive analysis, and the like. Using these, it is possible to appropriately rank piping systems and piping parts according to their FAC risks while ensuring objectivity.

The piping parts to be selected for monitoring include those with a high flow rate of cooling water, those with a cooling water temperature close to 130 to 150°C, those with a low concentration of oxidizer in the cooling water, and those with carbon steel pipes. Examples include piping parts formed with. As the piping system, a piping system is selected that includes such piping parts, has a long section outside the pressure vessel P12, and has a long section formed of carbon steel. For example, a purification system or a water supply system can be selected.

The piping parts selected for monitoring are the piping that constitutes the purification system, especially the heating side that connects the first stage heat exchanger P25a and the second stage heat exchanger P25b of the regenerative heat exchanger P25. communication pipe P42a on the heating side that connects the second stage heat exchanger P25b and the third stage heat exchanger P25c, and the second stage heat exchanger P25b and the first stage heat exchanger P42b. A communication pipe P42f on the cooling side that connects to the second heat exchanger P25a is preferable.

After executing the selection step S1, it is possible to proceed to any of steps S2 to S5. However, normally, in order to evaluate the FAC speed of the monitored object, the process moves to monitoring step S2. In steps S2 to S5, ECP monitoring and FAC measures are performed for the latest piping system and piping site selected in step S1.

(Monitoring step S2)
In the monitoring step S2, the performance index of the piping part selected as the monitoring target is monitored, and the corrosion rate (FAC rate) of flow accelerated corrosion (FAC) of the piping part estimated based on the performance index is monitored. Examples of the performance index include corrosion potential (ECP). Performance indicators such as ECP to be monitored are constantly monitored during the service life of the plant. FAC speed is estimated based on performance indicators such as ECP.

The ECP of a piping site can be measured by a corrosion potential sensor. As a corrosion potential sensor, use an appropriate electrode type sensor as long as it has the required heat resistance and pressure resistance, is equipped with a reference electrode that generates a reference potential, and can operate within the operating temperature range of the plant. be able to. The corrosion potential sensor can be installed at an appropriate location depending on the piping site selected as the monitoring target.

ECP monitoring may be performed using an existing corrosion potential sensor installed near the piping section selected as the monitoring target, or with a new corrosion potential sensor installed near the piping section selected as the monitoring target. A sensor may also be used. The corrosion potential sensor can be attached to a flange shaped like a pipe joint, a T-shaped pipe, or a guide port formed in the peripheral wall of the pipe. A flange or T-shaped pipe can be connected in the middle of a piping system to which a piping section selected as a monitoring target belongs.

Since the FAC rate has an electrochemical corrosion aspect, it can be expressed as a function of the corrosion current at the corrosion potential (ECP) of the material. If the correlation between the FAC speed and the ECP is determined for each piping site, the FAC speed can be estimated by applying the ECP monitoring results to the correlation.

The correlation between the FAC speed and ECP can be determined by actually measuring the wall thickness and ECP of a piping portion selected as a monitoring target. Although the ECP of the piping portion selected as a monitoring target is constantly monitored, the wall thickness of the piping portion can be measured intermittently through inspection with actual measurements. ECP is an important parameter related to FAC speed, so if ECP is constantly monitored at least, FAC speed can be estimated with high accuracy.

Since the FAC rate involves mass transfer, it can also be estimated using other parameters that affect FAC. Other parameters that affect FAC may be used to calibrate the correlation between FAC velocity and ECP, or may be used as input for a hydrodynamic analysis that takes mass transfer into account. The correlation between FAC speed and ECP can also be determined by multivariate regression using one or more of the parameters that affect FAC. Hydrodynamic analysis can be performed using as input one or more of the parameters that affect FAC.

For example, in fluid dynamic analysis, it is possible to construct a two-dimensional calculation system representing a cross section of a piping portion selected as a monitoring target. In such a calculation system, a basic equation is calculated using one or more of the parameters that affect FAC as input. Examples of basic equations include model equations expressing mass transfer and model equations expressing hydrodynamic behavior. In hydrodynamic analysis, it is also possible to take into account the effects of erosion, corrosion, droplet collision erosion, and the like.

By solving these basic equations simultaneously, the mass transfer and FAC velocity on the surface of the material can be estimated. The mass transfer analysis results reflect the electrochemical properties of the surface of the material, and therefore can be verified by comparison with ECP. By constantly monitoring ECP and using other parameters that affect FAC, it is possible to ensure the validity of the FAC speed estimated based on ECP.

The FAC speed is estimated using ECP and at least one of the following: oxygen concentration of cooling water, hydrogen peroxide concentration of cooling water, electrical conductivity of cooling water, iron ion concentration of cooling water, and pH of cooling water. It is preferable to do so. The oxygen concentration and hydrogen peroxide concentration of the cooling water affect the stability of the oxide film and the solubility of iron that makes up the piping parts. In addition, the electrical conductivity of the cooling water is related to the concentration of iron ions, etc., and is constantly monitored by installed equipment. Therefore, when these parameters are used together, the FAC speed can be estimated with high accuracy within a reasonable inspection cost.

One of the features of the method for improving the reliability of a nuclear power plant according to this embodiment is that a performance index such as ECP of a piping section is used as a constant monitoring index. Considering the state in which the FAC does not affect the operation of the BWR as the "desired performance" of the BWR with respect to the FAC, and demonstrating through constant monitoring that performance indicators such as ECP are in a state suitable for suppressing the FAC, Based on the definition that is important for realizing ER.

FAC speed is influenced by the corrosion resistance of the material. However, the corrosion resistance of materials cannot be constantly monitored during plant operation, and must be analyzed during plant construction or shutdown. Therefore, in the method for improving the reliability of a nuclear power plant according to this embodiment, the chemical composition of the material, such as the amount of Cr, is not used as a constant monitoring index. However, the chemical composition of the material, such as the amount of Cr, can be measured intermittently through inspections that involve actual measurement.

In the monitoring step S2, it can be confirmed whether the FAC speed estimated based on performance indicators such as ECP is within an acceptable target range for preventive maintenance. The estimated FAC speed based on the performance index can be compared to a preset threshold that defines the upper limit of a target range of FAC speeds. Any numerical value can be set as the threshold value depending on the magnitude of the corrosion allowance provided in the piping part to be monitored, the length of the plant's service period, and the like. The magnitude of the corrosion allowance is set based on the position of the piping portion, the material of the piping portion, the acceptability of the FAC, and the like.

After executing the monitoring step S2, it is possible to proceed to any of steps S3 to S6 depending on the result of comparing the FAC speed estimated based on the performance index with a threshold value.

For example, if the FAC speed is below the threshold, the FAC speed is within the target range, so the process can proceed to life cycle management step S6. On the other hand, if the FAC speed exceeds the threshold value, the FAC speed does not fall within the target range, so it is possible to proceed to correction step S4.

Alternatively, when the FAC speed exceeds the threshold value and it is predicted that the FAC speed will not fall within the target range even if correction step S4 is performed, it is possible to move to improvement step S5. Further, when the FAC speed exceeds the threshold value, and it is predicted that the FAC speed will not fall within the target range even if corrective step S4 is performed, and preventive maintenance measures have not been executed, preventive maintenance execution step S3 is performed. can be migrated.

(Preventive maintenance execution step S3)
In the preventive maintenance execution step S3, preventive maintenance measures are executed to reduce the flow accelerated corrosion (FAC) corrosion rate of the piping portion selected as the monitoring target. As preventive maintenance measures, oxygen injection, hydrogen peroxide injection, and titanium oxide injection, which are techniques to alleviate the corrosive environment of FAC, will be carried out. Further, in the preventive maintenance execution step S3, standard inspections can be performed on the piping portion selected as the monitoring target and piping portions other than the piping portion.

Oxygen injection, hydrogen peroxide injection, and titanium oxide injection can be performed for each piping system. In the nuclear power plant P100 shown in FIG. 2, an oxygen injection device P43 is connected to the purification system piping P20. The oxygen injection by the oxygen injection device P43 serves as a preventive maintenance measure for the piping portions belonging to the purification system. In fact, the oxygen injection device is used to protect the water supply pipe P10, and is installed in the same condensate system as the hydrogen injection device P16. A hydrogen peroxide injection device or a titanium oxide injection device can also be connected to the purification system piping P20 and the water supply piping P10.

Preventative maintenance measures can include oxygen injection, hydrogen peroxide injection, titanium oxide injection, and combinations of one or more of these. Hydrogen peroxide and titanium oxide solutions are prepared as liquids, so they do not require gas cylinders or gas piping, and are easy to handle and route around. Also, unlike gas, it is easy to increase the pressure, so it can be injected into each system using a small pump.

The amount of oxygen gas injected, the amount of aqueous hydrogen peroxide solution injected, and the amount of titanium oxide injected are set so that the FAC speed of the piping part selected for monitoring falls within an acceptable target range for preventive maintenance. can do. The behavior of the FAC velocity with respect to these injection amounts can be confirmed in advance by actually measuring the wall thickness of the piping portion or by performing a fluid dynamic analysis that takes into account mass transfer.

When oxygen or hydrogen peroxide is injected, the oxygen concentration or hydrogen peroxide concentration of the cooling water can be increased in the piping system to which the piping section selected as the monitoring target belongs. Since an oxide film is likely to be formed on the surface of the piping portion belonging to the piping system, the FAC speed can be reduced in the piping portion selected as the monitoring target and the piping portion in the vicinity thereof. In addition, when titanium oxide is implanted, the solubility of the oxide film decreases, so the FAC speed can be decreased. By performing these injections, it is possible to ensure the soundness of not only the piping portion selected as a monitoring target but also the piping portions in the vicinity thereof with respect to the FAC.

In the preventive maintenance execution step S3, standard inspections involving actual measurements can be performed on the piping parts selected as monitoring targets and piping parts other than the piping parts. Standard tests are performed intermittently as needed, as opposed to constant monitoring. A standard inspection is to measure the wall thickness of the pipe. The wall thickness can be measured using, for example, an ultrasonic thickness gauge. The composition of materials is usually managed in the material composition list at the time of plant construction. However, if a problem occurs due to FAC, the chemical composition of the material and oxide film will also be analyzed. The chemical composition can be measured by, for example, X-ray Fluorescence (XRF) analysis. Examples of the chemical composition include the amount of Cr in the base material, and the amount of Fe and O in the oxide film.

Test results of standard tests can be accumulated as data for each test in the process of repeating intermittent tests. The data accumulated through the repeated process can be compiled into a database as time-series data for each piping section. A database of wall thickness test results can be used to validate estimated FAC speeds based on performance metrics. A database of test results for the chemical composition of materials can be used to calibrate the correlation between FAC velocity and ECP and as input for hydrodynamic analysis.

According to the preventive maintenance execution step S3, preventive measures can be taken to prevent the occurrence and progression of FAC for the piping portion selected as the monitoring target. After executing the preventive maintenance execution step S3, it is possible to move to the monitoring step S2. In the monitoring step S2, the FAC speed after the preventive maintenance execution step S3 can be checked.

(Correction step S4)
In the corrective step S4, corrective measures are executed to control the water quality parameters of the cooling water flowing through the piping portion selected as the monitoring target. In this step, among the parameters that affect FAC, a water quality parameter related to the quality of cooling water is controlled to reduce the FAC speed of the piping portion selected as a monitoring target. It is preferable that the corrective step S4 is performed after the preventive maintenance execution step S3 is performed and the monitoring step S2 is performed.

Water quality parameters include oxygen concentration of cooling water, hydrogen peroxide concentration of cooling water, electrical conductivity of cooling water, pH of cooling water, concentration of dissolved hydrogen in cooling water, and concentration of impurities such as iron ions in cooling water. . As water quality parameters, at least one of the oxygen concentration of the cooling water, the hydrogen peroxide concentration of the cooling water, and the electrical conductivity of the cooling water is controlled because it has a large influence on FAC and is highly controllable. It is preferable to do so.

The oxygen concentration and hydrogen peroxide concentration in the cooling water can be changed by increasing the amount of oxygen gas injected into the cooling water, increasing the amount of hydrogen peroxide aqueous solution injected, or decreasing the amount of hydrogen injected within a range that does not affect SCC. can be increased by In addition, the oxygen concentration and hydrogen peroxide concentration of the cooling water should be adjusted by reducing the amount of oxygen gas injected into the cooling water, reducing the amount of hydrogen peroxide aqueous solution injected, and increasing the amount of hydrogen within a range that does not affect radioactive behavior. It can be lowered by increasing the injection volume.

Further, the oxygen concentration and hydrogen peroxide concentration of the cooling water can be controlled by operating the reactor within a range that does not affect the operation. For example, it can be increased by reducing the output of the reactor, increasing the amount of cooling water in the reactor, or lowering the temperature of the cooling water in the reactor. Further, it can be reduced by increasing the output of the reactor, decreasing the amount of cooling water in the reactor, or increasing the temperature of the cooling water in the reactor.

The electrical conductivity of the cooling water can be controlled by adjusting the concentration of impurities and the amount of titanium oxide injected. Examples of impurities include metal ions such as iron ions, chloride ions, sulfate ions, and the like. The concentration of impurities can be analyzed off-line by sampling through a sampling line or the like. The concentration of impurities can be controlled by adjusting the purification capacity of the condensate purification device, reactor water purification device, etc.

In the correction step S4, the water quality parameters of the cooling water are controlled so that the FAC speed estimated based on the ECP falls within an acceptable target range for preventive maintenance. In a correction step S4, the FAC rate corrected by controlling the water quality parameters can be compared with a preset threshold value. As the threshold value, an arbitrary numerical value can be set as in the monitoring step S2.

According to the correction step S4, highly responsive countermeasures can be taken to immediately suppress the occurrence and progression of FAC for the piping portion selected as the monitoring target. After performing the corrective step S4, a transition can be made to a monitoring step S2 in order to evaluate the monitored FAC speed. Thereafter, depending on the result of the comparison between the FAC rate corrected by controlling the water quality parameters and the threshold value, it is possible to proceed to one of steps S5-S6.

For example, if the FAC rate corrected by controlling the water quality parameters exceeds a threshold value, a return can be made to the correction step S4. In the re-correction step S4, the control details of the water quality parameters can be changed so that the FAC speed falls within an acceptable target range for preventive maintenance.

As a result of the change in the control content, if the FAC rate corrected by controlling the water quality parameters is below the threshold value, the FAC rate falls within the target range, so the process can proceed to life cycle management step S6. On the other hand, if the FAC rate corrected by controlling the water quality parameters exceeds the threshold value, the FAC rate does not fall within the target range, so the process can proceed to improvement step S5.

When water quality parameters are controlled, the oxygen concentration and hydrogen peroxide concentration of the cooling water increase, and the effect of suppressing stress corrosion cracking (SCC) by hydrogen injection or noble metal injection becomes weaker. Therefore, there is an upper limit to the increase in the amount of oxygen gas injected or the amount of hydrogen peroxide aqueous solution injected from the viewpoint of corrosion prevention maintenance. Furthermore, there are limits to the scope of control through nuclear reactor operation management and control of the electrical conductivity of cooling water. Therefore, if the FAC speed does not fall within the target range even after changing the control details of the water quality parameters, the process moves to improvement step S5 and the material of the piping part is replaced.

In addition, in the correction step S4, the water quality parameters of the cooling water are controlled so that the FAC speed estimated based on the performance index falls within an acceptable target range for preventive maintenance. Instead of speed, ECP can also be used as an index. When ECP is used as an index, the water quality parameters of the cooling water can be controlled so that ECP falls within a predetermined target range.

(Improvement step S5)
In the improvement step S5, improvement measures are performed to replace the material of the piping portion selected as the monitoring target. Materials for piping parts should be replaced if the corrosion rate of flow accelerated corrosion (FAC) does not fall within the target range even after corrective step S4, or if continuous measures are required after corrective step S4. can be done. Replacement of material in piping sections can also be suggested based on FAC velocity monitoring results by the computer managing the ER process for the FAC.

The material of the piping portion can be replaced with an appropriate material depending on the required purpose, material cost, etc., as long as the integrity of the FAC is ensured. However, the longer the piping section to be replaced, the higher the material cost will be. Additionally, materials with higher corrosion resistance are generally more expensive. Furthermore, weldability and the necessity of welding inspection vary depending on the material, which is related to maintenance costs. Therefore, it is preferable to select an appropriate material depending on the position and length of the piping portion.

The material of the piping section can be changed to the same type of material as the current piping section, as long as there is no substantial thinning due to corrosion. It is also possible to change the material to a material with higher corrosion resistance than the current piping parts. From the viewpoint of ensuring long-term soundness, it is preferable to change the material of the piping portion to a material with high corrosion resistance. The material of the piping portion may be changed to a different metal type, or may be changed to a different chemical composition of the same metal type.

The material of the piping part is carbon steel to which Cr is added at an impurity level of 0.3% by mass or less, low alloy steel to which Cr is added at 0.3% by mass or more and several mass% or less, stainless steel, etc. Can be exchanged. Low alloy steel is alloy steel in which the total amount of alloying elements is 5% by mass or less. Examples of low alloy steels include alloy steels containing about 0.3 to several mass % of Cr, 0.3 mass % or more of Ni, 0.08 mass % or more of Mo, and the like. Examples of the stainless steel include low carbon stainless steel containing 0.03% by mass or less of C.

According to the improvement step S5, it is possible to take continuous measures to suppress the occurrence and progression of FAC in the long term for the piping portion selected as the monitoring target. After the improvement step S5 is executed, it is possible to proceed to any one of steps S2 to S3 and S6. In the monitoring step S2, the FAC speed after execution of the improvement step S5 can be confirmed. In the preventive maintenance execution step S3, preventive maintenance measures can be implemented on the material after the improvement step S5 has been performed.

(Life cycle management step S6)
In the life cycle management step S6, the corrosion rate of flow accelerated corrosion (FAC) of piping in a nuclear power plant is managed by repeating the selection step S1, the monitoring step S2, the preventive maintenance execution step S3, the correction step S4, and the improvement step S5. . The content of the life cycle management step S6 is to repeat at least a portion of steps S1 to S5.

In life cycle management step S6, piping parts to be monitored are updated in accordance with the FAC risk each time steps S1 to S5 are repeated. Then, for the piping portion selected as a monitoring target, measurement of performance indicators such as ECP, estimation of FAC speed based on the performance indicators, and measures for FAC based on the estimated FAC speed are repeated. ER for FAC of piping in a nuclear power plant is realized by selecting measures for FAC and determining whether it is necessary to implement measures for FAC.

In the life cycle management step S6, the inspection conditions for the standard inspection performed in the preventive maintenance execution step S3 can be changed based on the FAC speed estimated based on the performance index. The change in inspection conditions is applied to the monitoring target selected in the selection step S1 and conventional periodic inspection targets other than the monitoring target. Inspection conditions include the frequency of inspection, the number of piping sections to be inspected, and the range of piping sections to be inspected.

In the process of repeating steps S1 to S5, in the monitoring step S2, if the FAC speed is below the threshold, the time interval for inspection may be extended, the number of piping parts to be inspected may be reduced, or the piping to be inspected may be The range of parts can be reduced. By changing the inspection conditions in this way, inspections involving actual measurements can be streamlined throughout the plant.

The measurement results of performance indicators such as ECP, the estimation results of FAC speed estimated based on the performance indicators, and the inspection results of standard inspections are collected as data for each piping part in the process of repeating steps S1 to S5. Can be accumulated. The data accumulated through the repeated process can be compiled into a database as time-series data for each piping section. Data representing performance index measurement results, data representing FAC speed estimation results, and data representing wall thickness inspection results actually measured for piping portions in standard inspections can be accumulated for each nuclear power plant.

In the life cycle management step S6, data representing the estimation results of the FAC speed accumulated in the process of repeating steps S1 to S5 and the actual wall thickness inspection results of the piping parts accumulated in the process of repeating steps S1 to S5 are used. A maintenance management database can be constructed for each nuclear power plant and each piping site based on the data representing the FAC speed and at least one of the data representing the FAC speed collected at other plants.

The maintenance management database can be used to verify the estimated FAC speed based on performance indicators. In the verification, the estimated amount of wall thinning based on the FAC speed estimation results collected at the nuclear power plant is compared with the actual amount of wall thinning based on the wall thickness inspection results collected at the nuclear power plant. Alternatively, the amount of thinning based on the FAC speed collected at the nuclear power plant is compared with the amount of thinning based on the FAC speed collected at other nuclear plants.

When comparing nuclear power plants with each other, it is preferable to compare data between piping parts whose parameters that affect FAC are similar between nuclear plants, as the amount of wall thinning based on FAC speed. It is preferable that the amount of thinning based on the FAC speed be verified for each nuclear power plant and for each piping site by comparing it with the actually measured amount of thinning.

As a result of the comparison, if the amounts of thinning match within a predetermined margin, it can be determined that the amount of thinning of the pipe selected as a monitoring target in the nuclear power plant is likely. In such a case, it is possible to record that the amounts of thinning match each other each time steps S1 to S5 are repeated. If it is confirmed that the thickness reduction amounts match each other over a predetermined cycle in repeating steps S1 to S5, the inspection conditions of the standard inspection performed in the preventive maintenance execution step S3 can be changed.

By changing inspection conditions based on such a maintenance management database, inspections involving actual measurements can be streamlined. In comparison with the case where the FAC speed estimated based on the performance index is compared with the threshold value in the monitoring step S2, the determination is made over a predetermined cycle. Therefore, the validity of rationalizing the inspection can be guaranteed to a higher degree.

According to the life cycle management step S6, long-term maintenance management of the FAC can be performed for the piping of the entire plant. In the life cycle management step S6, on the premise of constant monitoring of the performance index of the piping part to be monitored, the inspection time interval is determined for the piping part to be monitored and other piping parts belonging to the same piping system as the piping part to be monitored. It becomes possible to extend the period, reduce the frequency of examinations, and reduce the number of parts to be examined. The life cycle management step S6 can continue during the service life of the nuclear plant.

According to the ER process for FAC carried out in the method for improving the reliability of a nuclear power plant according to the present embodiment, unlike conventional technology, performance indicators such as ECP are constantly checked to ensure that the desired performance of the plant is not impaired. This can be ensured by monitoring. Since monitoring targets are selected according to FAC risk, ER can be implemented not only for piping parts selected as monitoring targets but also for piping systems other than monitoring targets. It is possible to reduce the amount of test items without reducing safety due to uncertainty. Therefore, such an ER process can simultaneously improve device reliability and streamline device maintenance. It becomes possible to review periodic inspection items and ensure the safety of workers in radiation exposure environments.

In conventional water quality management related to FAC, dissolved oxygen concentration, pH, and iron concentration are used as the main indicators. There is a problem that it is not possible to determine whether the location is upstream or not. On the other hand, if ECP is used as an index, the local corrosive environment where the sensor is installed is measured, so it is possible to precisely know the location where FAC occurs. The ECP of the piping material represents the corrosive environment of the cooling water flowing through the piping. Therefore, the corrosion potential sensor installed as a countermeasure against SCC can also be used as a countermeasure against FAC. ECP varies under the influence of pH. In addition, since it is affected by mass transfer determined by flow conditions, it fluctuates depending on the flow rate of cooling water, pipe diameter, geometric shape of the pipe, and temperature of cooling water. Therefore, ECP is useful as an index for precisely determining the location where FAC occurs.

<<Specific example of ER process for FAC>>
Next, a specific example of the ER process for FAC performed in the method for improving the reliability of a nuclear power plant according to the present embodiment will be described. FIG. 4 is a diagram showing a specific example of a flow of a method for improving reliability of a nuclear power plant according to an embodiment of the present invention.

As shown in I of FIG. 4, in the selection step S1, piping parts with a high risk of FAC are selected. Selection of piping parts may be performed for each plant or for each piping system. For example, if there are piping parts A to H, these piping parts are ranked according to their FAC risks. Piping section A with the highest risk of FAC is selected as the monitoring target for constant ECP monitoring.

As shown in II of FIG. 4, in the monitoring step S2, the ECP of the piping part selected as the monitoring target is monitored, and the FAC speed of the piping part estimated based on the ECP is monitored. The FAC speed to be monitored needs to be controlled so as to fall within a predetermined performance target range during the operation cycle from startup to shutdown of the plant. The performance target range can be set in advance for each operating cycle of the plant. It is confirmed that the FAC speed to be monitored matches the inspection data of the wall thickness which is intermittently measured through an inspection involving actual measurement.

As shown in III of FIG. 4, one of the preventive maintenance execution step S3, the correction step S4, and the improvement step S5 is executed depending on the FAC speed of the piping portion estimated based on the ECP. In the preventive maintenance execution step S3, for example, oxygen injection is performed. In the correction step S4, water quality parameters are controlled. In the improvement step S5, the material of the piping portion is replaced.

When oxygen is injected in the preventive maintenance execution step S3, the FAC speed decreases in the piping parts belonging to the piping system where oxygen is injected. However, the degree of reduction in FAC speed varies depending on the material of the piping, the temperature, the flow state of the cooling water, the geometrical shape of the piping, and the like. For example, for piping site A selected as a monitoring target, water quality parameters are controlled in correction step S4 so that the FAC speed falls within the target range.

In the preventive maintenance execution step S3 and the corrective step S4, the FAC speed may not fall to the target range. For example, piping parts B to E were ranked as high risk in the selection step S1, but the FAC speed did not fall to the target range. For such piping parts, the material of the piping parts is replaced in an improvement step S5 in order to continuously improve equipment reliability. Note that whether or not the FAC speed will fall to the target range may be determined based on the FAC speed estimated based on the ECP, or based on past inspection results and operating experience at other plants. Good too.

In life cycle management step S6, in response to changes in the FAC risk ranking, piping parts are reselected, the ECP to be monitored is measured, the FAC speed is estimated based on the ECP measurement results, and measures are taken against the FAC. repeat. Such an ER process turns FAC risk into ALARP. Since the FAC risk is reduced for the piping site selected as a monitoring target, it is possible to clearly indicate that the FAC risk is reduced for other piping sites that belong to the same piping system as the piping site. Therefore, the long-term reliability of nuclear power plants can be improved.

<<Example of application of ER process for FAC to BWR>>
Next, an example of application of the ER process for FAC to BWR, which is performed in the method for improving reliability of a nuclear power plant according to the present embodiment, will be described. FIG. 5 is a diagram showing an example of application of the flow of the method for improving reliability of a nuclear power plant according to the embodiment of the present invention to BWR.

As shown in I of FIG. 5, in BWR, the FAC speed differs for each piping section. In the selection step S1, piping parts with a high risk of FAC are selected from the reactor coolant purification system and water supply system as candidates. These piping systems have a long section outside the pressure vessel and often use carbon steel pipes, so they are piping systems with a high risk of FAC. Possible piping parts include communication pipes of regenerative heat exchangers, purification system piping, water supply piping, and the like.

Each stage of heat exchanger includes a tube-side communication pipe and a shell-side communication pipe. For example, in the case of a three-stage heat exchanger, there are a total of four candidates on the tube side and the shell side. For example, the communication pipe with the highest risk of FAC is selected as the monitoring target for constant ECP monitoring. Piping parts predicted to have a low risk of FAC can be excluded from selection candidates in advance. For example, it is predicted that the risk of FAC is low in piping parts where the cooling water is at a low temperature.

As shown in II of FIG. 5, when the preventive maintenance execution step S3 and the corrective step S4 are executed, the FAC speed decreases for each piping section. For example, the connecting pipe selected in the selection step S1 is controlled so that the corrosion rate of the FAC is within the target range. If the FAC speed does not fall within the target range due to oxygen injection, a reduction in the amount of hydrogen injection or an increase in the amount of noble metal injection can be combined within a range that does not affect the SCC. However, like piping part B, there are piping parts where the FAC speed is difficult to decrease depending on the material of the piping, the temperature, the state of the flow of cooling water, the geometrical shape of the piping, etc.

As shown in III of FIG. 5, when the improvement step S5 is executed, the FAC speed is significantly reduced in the piping portion where the material has been replaced. For example, a piping part where the FAC speed is unlikely to decrease, such as piping part B, is selected as a monitoring target in the next cycle and thereafter, and the material is replaced in the improvement step S5. In addition, it is difficult to inject oxygen into piping on the upstream side and downstream side of the purification system, and piping on the downstream side of the water supply system, such as the bottom drain piping P34, because the cooling water is at a high temperature. Such piping parts are dealt with by replacing the material in the improvement step S5.

As shown in IV of FIG. 5, when the life cycle management step S6 is executed, the FAC risk becomes ALARP for any piping site. For example, connecting pipes have been the piping site with the highest risk of FAC, but various measures can reduce the risk of FAC. By continuing to clearly demonstrate through constant ECP monitoring that the FAC risk of connecting pipes is decreasing, it becomes possible to demonstrate that the FAC risk is decreasing for piping parts other than connecting pipes as well. .

Therefore, when the life cycle management step S6 is executed, on the premise that the ECP of the piping part to be monitored is constantly monitored, inspections of the piping part to be monitored and other piping parts belonging to the same piping system as the piping part to be monitored are performed. It is possible to extend the time interval, reduce the frequency of examinations, and reduce the number of parts to be examined. It is possible to streamline equipment maintenance such as inspections while ensuring the safety of workers in radiation exposure environments and equipment reliability.

≪Modification 1≫
In the monitoring step S2, auxiliary performance indicators can be constantly monitored in addition to performance indicators such as ECP of the piping portion. The auxiliary performance indicators include at least one of the oxygen concentration of the cooling water, the hydrogen peroxide concentration of the cooling water, the electrical conductivity of the cooling water, the iron ion concentration of the cooling water, and the pH of the cooling water. Another example is the amount of a substance deposited on the surface of the material of the piping portion by noble metal injection, titanium oxide injection, or the like.

These supplementary performance indicators are periodically correlated with FAC speed. Based on such a correlation, the FAC velocity of the piping section can be estimated. By constantly monitoring the auxiliary performance index, equipment reliability can be guaranteed even in piping parts where a corrosion potential sensor cannot be installed or during periods where ECP cannot be measured due to failure of the corrosion potential sensor.

Note that the correlation between the auxiliary performance index and the FAC speed may be affected by the output of the reactor, and may change depending on the burnup of the nuclear fuel, the position of the control rods, the amount of cooling water in the reactor core, etc. Therefore, the correlation between auxiliary performance indicators and FAC speed is difficult to maintain in the long term. Therefore, it is preferable that such a correlation is periodically determined and updated every time the reactor fuel is replaced or every predetermined operating time. For example, it is preferable to update at least every five years or so when all the fuel in the reactor is replaced.

The oxygen concentration of the cooling water and the hydrogen peroxide concentration of the cooling water can be increased until the FAC speed decreases as long as the effect of suppressing SCC by hydrogen injection or noble metal injection can be obtained. On the other hand, noble metal injection catalyzes the recombination reaction and maintains the ECP near the redox potential of hydrogen. Therefore, when using the amount of adhesion due to noble metal injection as an auxiliary performance index, it is preferable to confirm that hydrogen is present in a ratio greater than or equal to the stoichiometric ratio of water to oxygen. Within this range, the amount of deposition due to noble metal implantation becomes a highly reliable performance index.

Note that titanium oxide injection is a technique in which a solution of titanium oxide is injected into cooling water to cause titanium oxide to adhere to the inner surface of piping. Titanium oxide is known to exhibit photocatalytic activity under Cerenkov light and reduce the corrosion potential of materials. Titanium oxide is incorporated into the oxide film and produces ilmenite (FeTiO ₃ ) with low solubility, so that FAC can be suppressed.

The electrical conductivity of the cooling water and the iron ion concentration of the cooling water can be reduced until the FAC speed is reduced within the range of the quality of the cooling water and the purification ability of the cooling water. While monitoring ECP and FAC rates, it may be determined that an increase in FAC rate is due to changes in the water quality itself, rather than to countermeasures such as oxygen injection or ECP. In such a case, the purification ability for cooling water can be increased during the correction step S4 and other steps.

≪Modification 2≫
In the monitoring step S2, the wall thickness of the piping portion may be used instead of the ECP of the piping portion as a performance index that is constantly monitored. The wall thickness of piping parts can be directly measured during plant operation using an ultrasonic thickness gauge that is compatible with high-temperature environments. The wall thickness of the piping portion can be measured at one or more measurement points, including the piping portion selected as the monitoring target. However, unlike ECP, the wall thickness of the piping portion tends to cause measurement noise. Therefore, from the viewpoint of taking prompt measures against FAC and improving controllability of measures, it is preferable to constantly monitor ECP.

≪Modification 3≫
FIG. 6 is a diagram showing an example of a nuclear power plant to which the method for improving reliability of a nuclear power plant according to the embodiment of the present invention is applied. FIG. 6 illustrates an advanced boiling water reactor (ABWR). The method for improving the reliability of a nuclear power plant described above can also be applied to other types of nuclear power plants such as ABWR.
As shown in FIG. 6, the nuclear power plant P200 is an ABWR and includes a reactor P1, a turbine P3, a condenser P4, a reactor coolant purification system, a water supply system, and the like.

Nuclear power plant P200 is different from nuclear power plant P100 and does not include a recirculation system. Further, the nuclear power plant P200 includes an internal pump P40 instead of the jet pump P21. The downstream side of the bottom drain pipe P34 branches into a purification system pipe P20 and a residual heat removal system pipe P20a. The downstream side of the residual heat removal system piping P20a communicates with a downcomer P17 inside the pressure vessel P12.

Further, the turbine P3, the high-pressure feed water heater P9, and the low-pressure feed water heater P8 are connected to each other by an air extraction pipe P14. The extracted steam extracted from the steam bypasses the condenser P4 and is introduced into the high pressure feed water heater P9 and the low pressure feed water heater P8. The drain of the high-pressure feed water heater P9 is configured to be returned to the water supply. The drain of the low-pressure feedwater heater P8 is configured to be returned to condensate.

In the nuclear power plant P200, the corrosion potential sensor P35a is installed in a branch pipe P34a branched from the bottom drain pipe P34. Further, a corrosion potential sensor P35e is attached to a flange P36 connected to the residual heat removal system piping P20a. Corrosion potential sensors P35a and P35e are provided to confirm the effect of suppressing stress corrosion cracking (SCC) due to hydrogen injection or noble metal injection.

In the nuclear power plant P200, corrosion potential sensors for measuring the corrosion potential (ECP) of materials in contact with cooling water are installed in the communication pipes P42a, P42f, etc. of the regenerative heat exchanger P25, as in the nuclear power plant P100. be able to. Furthermore, an injection device that performs oxygen injection, hydrogen peroxide injection, and titanium oxide injection can be connected to the purification system piping P20 and the water supply piping P10.

Although the embodiments of the present invention have been described above, the present invention is not limited to the embodiments described above, and various changes can be made without departing from the spirit of the present invention. For example, the present invention is not necessarily limited to having all the configurations of the embodiments described above. Replacing part of the configuration of one embodiment with another configuration, adding part of the configuration of one embodiment to another form, or omitting part of the configuration of one embodiment Can be done.

S1 Selection step S2 Monitoring step S3 Preventive maintenance execution step S4 Correction step S5 Improvement step S6 Life cycle management step P100 Nuclear power plant P1 Reactor P2 Main steam piping P3 Turbine P4 Condenser P5 Condensate pump P6 Condensate purification device P7 Water supply pump P8 Low pressure feed water heater P9 High pressure feed water heater P10 Water supply piping P11 Containment vessel P12 Pressure vessel P13 Core P14 Bleed piping P15 Shroud P16 Hydrogen injection device P17 Downcomer P18 Hydrogen injection piping P19 Opening/closing valve P20 Purification system piping P21 Jet pump P23 Purification system isolation Valve P24 Purification system pump P25 Regenerative heat exchanger P26 Non-regenerative heat exchanger P27 Reactor water purification device P30 Recirculation system piping P31 Precious metal injection device P32 Precious metal injection piping P33 On-off valve P34 Bottom drain piping P35 Corrosion potential sensor P36 Flange P38 Neutron meter Pipe P40 Internal pump P41 Shroud support P42 Connecting pipe P43 Oxygen injection device P44 Oxygen injection pipe P45 Opening/closing valve

Claims (9)

a selection step of selecting piping parts with a high risk of flow-accelerated corrosion;
a monitoring step of monitoring the corrosion rate of flow-accelerated corrosion of the selected piping portion;
performing preventive maintenance to reduce the corrosion rate of the piping portion;
a corrective step of controlling water quality parameters of cooling water flowing through the piping section;
an improvement step of replacing the material of the selected piping section if the corrosion rate does not fall within the target range even after performing the correction step;
a life cycle management step of managing the corrosion rate of piping in a nuclear power plant through the selection step, the monitoring step, the preventive maintenance execution step, the correction step and the improvement step;
A method for improving the reliability of a nuclear power plant with The method for improving reliability of a nuclear power plant according to claim 1, comprising:
In the selection step, the piping parts are selected using at least one of a risk analysis tool and an anthropogenic risk analysis based on the importance classification regarding the structures, systems, and equipment of the nuclear power plant. How to improve sex. The method for improving reliability of a nuclear power plant according to claim 1, comprising:
A method for improving the reliability of a nuclear power plant, wherein in the monitoring step, a corrosion potential of the selected piping portion is monitored, and a corrosion rate of flow-accelerated corrosion of the piping portion estimated based on the corrosion potential is monitored. The method for improving reliability of a nuclear power plant according to claim 3,
In the monitoring step, at least one of the corrosion potential, the chemical composition of the piping section, the quality of the cooling water, the temperature of the cooling water, and the hydrodynamic characteristics of the cooling water in the piping section A method for improving the reliability of a nuclear power plant, estimating the corrosion rate based on the corrosion rate. The method for improving reliability of a nuclear power plant according to claim 1, comprising:
In the monitoring step, at least one of the oxygen concentration of the cooling water, the hydrogen peroxide concentration of the cooling water, and the electrical conductivity of the cooling water is measured, and the measurement result is used to calibrate the corrosion rate. How to improve plant reliability. The method for improving reliability of a nuclear power plant according to claim 1, comprising:
In the preventive maintenance execution step, at least one of oxygen injection, hydrogen peroxide injection, and titanium oxide is performed. A method for improving reliability of a nuclear power plant according to claim 1, comprising:
In the corrective step, at least one of increasing the amount of oxygen injected into the cooling water, increasing the amount of hydrogen peroxide injected into the cooling water, and decreasing the amount of hydrogen injected into the cooling water is implemented. Methods for improving reliability of nuclear power plants. The method for improving reliability of a nuclear power plant according to claim 1, comprising:
In the life cycle management step, the data representing the corrosion rate accumulated in the process of repeating the selection step, the monitoring step, the preventive maintenance execution step, the correction step, and the improvement step, and the actual measurement of the corrosion rate of the piping portion. A nuclear power plant that constructs a database for each nuclear power plant and each piping section based on at least one of data representing the thickness inspection results and data representing the corrosion rate of flow accelerated corrosion collected at other plants. How to improve plant reliability. The method for improving reliability of a nuclear power plant according to claim 8,
Based on the database, the time interval between wall thickness inspections for each piping section is extended, and the corrosion potential of each piping section is monitored during the extended period, thereby improving equipment reliability against flow-accelerated corrosion for each piping section. A method for improving the reliability of nuclear power plants to ensure

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